ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 24

 New Total Rewrite Revision Page Change

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Document Title:

Paducah Gaseous Diffusion Plant – Occupational 

Internal Dose

Document Number: ORAUT-TKBS-0019-5 

Revision: 01 PC-1 

Effective Date: 01/31/2007

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert(s): Daniel S. Mantooth and Clark B. Barton

Site Expert(s): N/A

Approval: Signature on File Approval Date: 01/11/2007

Daniel S. Mantooth, Document Owner

Approval: Signature on File Approval Date: 01/12/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/11/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 01/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 01/31/2007

James W. Neton, Associate Director for Science
Document No. ORAUT-TKBS-0019-5 Revision No. 01 PC-1 Effective Date: 01/31/2007 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

09/30/2004 00 New technical basis document for the Paducah Gaseous Diffusion 

Plant – Occupational Internal Dose. First approved issue. Initiated 

by Jay J. Maisler. 

10/25/2006 01 Approved issue of Revision 01. This revision addresses the 

comment from the meeting with the United Steelworkers Local 5-550 

and SPFPA Local 111 held on 2/10/2005. Constitutes a total rewrite 

of document. Incorporates recent direction from NIOSH to include 

DOL review comments on page 5. Revisions address corrective 

actions responding to OCAS assessment and internal review 

comments from Task 5. Changes were made at the end of Section 

5.2. The original Table 5-3 has been deleted and replaced with an 

expansion of the original Table 5-4, which becomes Table 5-3. Table 

5-2 has been updated to reflect new Table numbers. Incorporates 

internal and NIOSH formal review comments. This revision results in 

an increase in assigned dose and a PER is required. Training 

required: As determined by the Task Manager. Initiated by Daniel S. 

Mantooth. Approval:

Daniel S. Mantooth, Document Owner

Signature on File___________________________ ___10/12/2006 

John M. Byrne, Document Owner

Signature on File___________________________ ___10/12/2006 

Edward F. Maher, Task 5 Manager

Signature on File____________________________ __10/12/2006 

Kate Kimpan, Project Director

Signature on File________ ________________ 10/23/2006 

James W. Neton, Associate Director for Science

Brant A. Ulsh Signature on File for ______________ _ 10/25/2006 

01/31/2007 01 PC-1 Approved page change revision initiated to correct the title of Table 

5-2 on page 10. Modified Table 5-4 on page 13 to consolidate the 

years of 1952-1998 onto the first line of the table. Added Table 5-2a 

to the List of Tables. Incorporates formal internal and NIOSH review 

comments. Includes changes to Table 5-2a (page 11). This revision 

results in no change to the assigned dose and no PER is required. 

Training required: As determined by the Task Manager. Initiated by 

Daniel S. Mantooth. 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 3 of 24

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ............................................................................................................... 4

PAGE

5.1 Introduction .............................................................................................................................. 5

5.1.1 Purpose........................................................................................................................ 6

5.1.2 Scope........................................................................................................................... 6

5.2 Source Term ............................................................................................................................ 6

5.2.1 Isotopic Concentrations ................................................................................................ 7

5.3 In Vitro Measurement Methods .............................................................................................. 11

5.3.1 Measurement Types and Detection Levels ................................................................. 12

5.3.2 Reporting Formats and Codes....................................................................................12

5.4 In Vivo Measurement Methods............................................................................................... 12

5.4.1 Measurement Types and Detection Levels ................................................................. 13

5.4.2 Reporting Formats and Codes....................................................................................13

5.4.3 Instructions for Addressing Possible Interferences and Uncertainties......................... 13

5.4.4 Assessment of Intake for Monitored Employees .........................................................13

5.5 Significant Incidents with Internal Dose Potential ................................................................... 16

References ........................................................................................................................................19

Glossary ............................................................................................................................................21

LIST OF TABLES

TABLE TITLE

5-1 Radiological source term for PGDP processes and facilities ....................................................

PAGE

7

5-2 Bounding isotopic concentrations for PGDP operations (pCi/g U)a ......................................... 10

5-2a Bounding isotopic activity concentrations for PGDP operations (pCi/pCi U)........................... 11

5-3 In vitro measurement frequencies .......................................................................................... 12

5-4 In vitro measurement types and detection levels for various periods...................................... 13

5-5 In vitro record codes...............................................................................................................14

5-6 In vivo measurement types and detection levels for various periods ...................................... 16

5-7 In vivo record codes...............................................................................................................17

5-9 A history of significant incidents and events ........................................................................... 18
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 4 of 24

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

Bq becquerel

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

g gram

hr hour

KPA kinetic phosphorescence analysis

L liter

MDA minimum detectable amount

MDC minimum detectable concentration

mg milligram

mL milliliter

mo month

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

pCi picocurie

PGDP Paducah Gaseous Diffusion Plant

Ppb parts per billion

ppm parts per million

RU Recycled Uranium

TRU transuranic

U.S.C. United States Code

UF6 uranium hexafluoride

UO3 uranium trioxide

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 5 of 24

5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety (NIOSH) but are rather general working documents that 

provide historic background information and guidance to assist in the preparation of dose 

reconstructions at particular sites or categories of sites. They will be revised in the event additional 

relevant information is obtained about the affected site(s). These documents may be used to assist 

NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document, the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 6 of 24

5.1.1

This TBD provides technical data and other key information which will serve as the technical basis for 

evaluating internal occupational radiation dose for EEOICPA claimants who were employed at the 

Paducah Gaseous Diffusion Plant.

Purpose

5.1.2

This document discusses the radionuclides potentially encountered by Paducah Gaseous Diffusion 

Plant (PGDP) employees during the Plant’s operational history. The PGDP mission was to enrich 

uranium in the form of uranium hexafluoride (UF6) from roughly 0.7% 235U (natural enrichment) to 

around 3% 235U for use in domestic and foreign power reactors (DOE 2000, p. 8). Enrichment 

operations began in 1952 in the first four process buildings, C-331, C-333, C-310, and C-315. From 

1953 until 1977, UF6 feed material was produced from uranium trioxide (UO3) at the plant. From 1953 

to 1964, and again from 1968 to 1977, UF6 was produced from the recycled uranium (RU) produced 

from spent reactor fuel. In May 1977, the feed plants ceased operation and all feed to the enrichment 

process was in the form of UF6 obtained from outside sources. Other chemical compounds of 

uranium were present throughout the Plant’s history including UO2F2, UF4, and UO3. 

Scope

ORAUT (2004a) contains detailed information on the history of PGDP and the feed conversion and 

enrichment process. 

5.2 SOURCE TERM

The radionuclides potentially encountered by PGDP employees consist primarily of the isotopes of 

uranium, 238U, 234U, and 235U. The progeny of dosimetric interest for these radionuclides includes 

230Th and 234mPa (BJC 1999, p. 8). While a range of 235U enrichment values (from natural to 3%) was

encountered in feed and product material, the dose reconstructor should assume a nominal 

enrichment value of 2% 235U for all feed or product materials.

The presence of transuranic (TRU) and fission product isotopes has been known since the early days 

of operation (ORAUT 2004a). These radionuclides were present in small amounts in the feed 

material produced from RU, and include 237Np, 239Pu, and 241Am. Technetium-99 is the fission product 

of concern from a dosimetry standpoint. During the conversion of the feed material from UO3 to UF6, 

the elemental species of the TRU and fission products react differently from a chemical standpoint. 

For example, 99Tc tends to form a very volatile fluoride; readily introduced into the enrichment 

cascade, it essentially follows the UF6 to the surge and/or product station. On the other hand, both 

241Am and 239Pu form essentially nonvolatile fluorides. Most of these isotopes will remain in the feed 

conversion byproducts (i.e., ash) or in the feed cylinder heels, a primary source of TRU exposure from 

this process. Some 241Am will appear in the cascade components as a result of the decay of 241Pu. 

While 237Np forms a volatile fluoride, it rapidly oxidizes and plates on the metal cylinder walls (PACE 

and University of Utah 2000, p. 26), so the main hazard from this isotope is related to cylinder 

cleaning operations and cascade component maintenance. PGDP conducted operations to recover 

99Tc (April 1960 through June 1963) and 237Np (November 1958 through March 1962), which presents 

an increased potential to exposure from these isotopes.

Table 5-1 lists information on the isotopic species associated with PGDP operations and facilities and 

their dates of operation (when available).
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 7 of 24

5.2.1

At PGDP, monitoring for intakes of uranium, in vivo or in vitro, often resulted in reports of elemental 

uranium concentration in urine or the mass of elemental uranium in organs or the whole 

Isotopic Concentrations

Table 5-1. Radiological source term for PGDP processes and facilities. 

Building Process Datesa 

Radionuclides 

of concernb

Absorption 

typec

C-360 Product shipping and transfer U-234, -235, -236, -238 F

Np-237 M

Pu-238, -239, -240, -242 M, S

Tc-99 F

C-400 Converter maintenance 4/53–Present U-234, -235, -236, -238 F

Converter salvage Np-237 M

CIP/CUP Phase 1 9/54–6/61 Pu-238, -239, -240, -242 M, S

CIP/CUP Phase 2 3/73–9/81 99-Tc F

C-400/ 237Np recovery 11/58–3/62 U-234, -235, -236, -238 F

C-710 99Tc recovery 4/60–6/63 Np-237 M

Pu-238, -239, -240, -242 M, S

99-Tc F

C-409 Converter refurbish 1973–1981 U-234, -235, -236, -238 F

Np-237 M

Pu-238, -239, -240, -242 M, S

Tc-99 F

C-410 Green salt production 1953–1956 U-234, -235, -236, -238 F, M, S

UF6 production 1953–Present Np-237 M

99Tc /237Np recovery (MgF2) 9/61–6/63 Pu-238, -239, -240, -242 M, S

99-Tc F

Th-230 S

Am-241 M

C-420 Green Salt Production 1956–1977 U-234, -235, -236, -238 F, M, S

Np-237 M

Pu-238, -239, -240, -242 M, S

Tc-99 F

Th-230 S

Am-241 M

C-331 Cascade ops/maintenance 9/52–Present U-234, -235, -236, -238 F

C-333 Cascade ops/maintenance 9/52–Present Np-237 M

C-335 Cascade ops/maintenance 4/54–Present Pu-238, -239, -240, -242 M, S

C-337 Cascade ops/maintenance 7/54–Present Tc-99 F

C-310 Surge and product 9/52–Present U-234, 235, 236, 238 F

99Tc /237Np recovery (MgF2) 1/63–6/63 Np-237 M

Pu-238, -239, -240, -242 M, S

Tc-99 F

Th-230 S

C-315 Surge and waste 9/52–Present U-234, -235, -236, -238 F

Th-230 S

C-340 Metal production 12/57–12/62 U-234, -235, -236, -238 M, S

1/68–10/77 Np-237 M

Pu-238, -239, -240, -242 M, S

Th-230 S
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 8 of 24

Building Process Datesa Radionuclides of concernb,c

Absorption 

typeb,c 

C-720 Maintenance U-234, -235, -236, -238 F, M, S

Np-237 M

Pu-238, -239, -240, -242 M, S

Tc-99 F

Th-230 S

Am-241 M

C-746 Waste metal smelting U-234, -235, -236, -238 M, S

Np-237 M

Pu-238, -239, -240, -242 M, S

Th-230 S

C-749 Uranium metal burial U-234, -235, -236, -238 M, S

Np-237 M

Pu-238, -239, -240, -242 M, S

Th-230 S

a. All dates are from DOE (2000) except those for C-409 and the Cascade Ops/Maintenance buildings, which are from 

PACE and University of Utah (2000) and ORAUT (2004a), respectively.

b. From Hill and Strom (1993, Table 16.2).

c. From BJC (1999, Table 1.7). 

body. However, the internal dose assessment process requires the use of isotopic concentrations as 

input. As stated above, the chemical properties of the TRU and fission product contaminants resulted 

in changes in their relative concentrations in different parts of the process when compared to the 

original feed material. This section provides bounding estimates for radionuclide concentrations. The 

concentrations relative to the mass of total uranium and the activity of total uranium are shown in 

Tables 5-2 and 5-2a, respectively. A summary discussion on the information presented in the tables 

and its derivation is provided below. Details of units conversion and data sources can be found in 

Barton (2006).

Pre-1983 Operations

Column 1, Pulverizer, Ash, Green Salt C-410; Column2, Converter Salvage

The values presented in Columns 1 and 2 are taken from PACE and University of Utah (2000) Table 

7.9. The maximum air concentration data was used to provide bounding values for TRU 

concentrations. The data in Table 7.9 is in units of uCi/cc which were converted to the desired units 

of pCi (TRU)/g (U) or pCi (TRU)/pCi(U). Details of this process can be found in Barton (2006).

The maximum plutonium activity in air is reported in PACE and the University of Utah (2000) as being 

comprised of 239Pu. However, it is more consistent with all references (BJC[2000], Hightower et. al. 

[200]) to assume that the faction of the alpha activity attributed to 239Pu be considered as total alphaemitting isotopes plus 241Am. A nominal isotopic distribution for TRU in fuel-grade plutonium provided 

in ORAUT (2005) was used to calculate the ratios of the various isotopes from the value provided in 

the reference. 

Column 1 values are applicable to feed material production operations in C-410/C-420. Column 2 

values are applicable to converter maintenance/salvage and cylinder cleaning operations in C-400. 

Column 4, Tc/Np Recovery Operations; Column 5, Balance of Plant
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 9 of 24

The values in Column 4 and Column 5 are taken from BJC (2000), Table 2.4.1. This reference 

presents a detailed analysis of the mass balance of 239Pu, and 237Np using individual activity analysis. 

Table 2.4.1 presents the maximum concentrations (in ppb) of the 239Pu and 237Np for the various 

processes encountered at PGDP. The remaining plutonium isotopes and 241Am concentrations are

determined using the isotopic rations for fuel grade plutonium found it ORAUT (2005) and as 

described above.

The values in Column 4 are applicable to all Tc/Np recovery operations in C-400, including those 

that may have taken place after 1983.

The values for Column 5 consist of the upper 95% bound of the average Table 2.41 values (minus 

those areas/processes already accounted for in Columns 1 and 2). They are applicable to pre-1983 

activities which include Cascade Operations (C-333, C-337, C-410), Production/Handling UF4 (C-340), 

Connecting/Disconnecting UF6 tails Cylinders (C-315, C-340), Connecting/Disconnecting Product 

Cylinders (C-310), Changing/Cleaning MgF2 Traps (C-410, C-310), and U Metal Production (C-340) 

BJC (2000), Table 2.4.1.

Post-1983 Operations

Column 3, Converter Salvage; Column 6, Balance of Plant

Converter Salvage (Column 3) and converter salvage operations (Column 6) are based on Hightower 

et. al. (2000). These values were included in the TBD to account for the fact that an insignificantly 

small percentage (.008% of the total) of RU feed was introduced to the cascade after 1977 (BJC 

2000, Appendix C). In addition, a two-phase (1954 to 1961 and 1973 to 1981) upgrade program 

effectively replaced the major components of the cascade. These facts would imply that, over time, 

TRU and fission product contaminants would be greatly reduced from the process, with the exception 

of those entrained in the feed and tail cylinder heels. Only cylinder maintenance/washing operations 

would have a potential for exposure to these isotopes. Hightower et. al. (2000) contain analytical data 

in ppb for 237Np, 239Pu, and 241Am in depleted uranium cylinders and feed material. The values in 

Table 3, the bounding concentrations for UF6 feed, were used to represent Converter Salvage 

Operations (Column 3). As in the case for the Pre-1983 operations, these include converter 

maintenance/salvage and cylinder cleaning in C-400. The values for the remaining plutonium isotopes 

were determined as using the ratios in ORAUT (2005), as described above.

For Balance of Plant Operations (Column 6) 239Pu, 237Np, and 241Am values specified in Hightower et. 

al. (2000) Table 2 were used. These data represent the concentrations of dispersed contamination in 

DUF6 cylinders and, considering the negligible quantity of RU introduced to the plant during this 

period, is believed to bound the actual 239Pu and 237Np levels since these isotopes were primarily 

concentrated in the pulverizer ash or plated within feed cylinder walls. 

Column 6 is applicable to post-1983 activities which include Cascade Operations (C-333, C-337, C410), Production/Handling UF4 (C-340), Connecting/Disconnecting UF6 tails Cylinders (C-315, C-340), 

Connecting/Disconnecting Product Cylinders (C-310), Changing/Cleaning MgF2 Traps (C-410, C310), and U Metal Production (C-340) BJC (2000), Table 2.4.1.
Document No. ORAUT-TKBS-0019-5 Revision No. 01 PC-1 Effective Date: 01/31/2007 Page 10 of 24

Thorium 230 (230Th) 

Pace and University of Utah (2000) list 230Th, a decay product of uranium as a potential source of 

concern and provide estimates for the activity concentration in Buildings C-400 and C-410 (Columns 1 

and 2 of Table 5-2 and 5-2a). BJC (2000) and Hightower et. al. (2000) offer no data on 230Th. The 

ratio of Np-237 to Pu-239 data from each BJC (2000) Table 2.4.1 and Hightower et. al. (2000) was 

compared to the same ratio determined for the PACE “mixes”. The mix for which the two ratios 

compared most closely was used to determine the 230Th concentration. A detailed discussion of this 

technique is provided in Barton (2006).

Technetium 99 (99Tc)

The 99Tc concentration for all operations except 99Tc recovery is assumed to remain constant 

throughout the process and is based on the reported 7-ppb maximum for the feed material (Smith

1984). The 99Tc concentration for recovery operations is from BJC (2000).

Note: If the work location of a claimant is unknown, use Columns 2 or 3 in Table 5-2 (or Table 

5-2a), as appropriate to the times of employment.

 

Table 5-2. Bounding isotopic concentrations for PGDP operations (pCi/g 

U)a.

Radionuclide

1 2 3 4 5 6

Pulverizer, 

ash 

handling, 

green salt 

C-410 

Pre-1983b

Converter 

Salvage 

Line 

Pre-1983b

Converter 

Salvage 

Line 

Post1983c

Tc/Np 

recovery 

Operations, 

C-400d

Balance 

of plant 

Pre1983d

Balance of 

plant Post1983c

Np-237 3.55E+04 1.61E+06 3.81E+04 1.76E+07 1.67E+03 3.67E+00

Pu-239/240 9.00E+05 3.24E+04 7.09E+04 4.42E+06 4.11E+01 1.95E+00

Pu-238 1.94E+05 7.01E+03 1.53E+04 9.54E+05 8.89E+00 4.21E-01

Pu-242 4.46E+01 1.61E+00 3.52E+00 2.19E+02 2.04E-03 9.67E-05

Pu-241 3.51E+07 1.27E+06 2.76E+06 1.72E+08 1.61E+03 3.38E-01

Am-241 1.56E+05 5.62E+03 1.85E+03 7.65E+05 7.13E+00 4.21E+00

Th-230 1.18E+06 1.67E+05 3.64E+05 2.27E+07 2.11E+02 1.00E+01

Tc-99 1.20E+05 1.20E+05 1.20E+05 3.81E+07 1.20E+05 1.20E+05

U-234 6.81E+05 6.81E+05 6.81E+05 6.81E+05 6.81E+05 6.81E+05

U-235 4.27E+04 4.27E+04 4.27E+04 4.27E+04 4.27E+04 4.27E+04

U-238 3.27E+05 3.27E+05 3.27E+05 3.27E+05 3.27E+05 3.27E+05

U-236 9.00E+02 9.00E+02 9.00E+02 9.00E+02 9.00E+02 9.00E+02

a. Based on uranium isotopic distribution for 2% enriched U (BJC 2000, Table 1-5). Assuming enriched as opposed to 

natural or depleted uranium provides bounding values for TRU intake estimates.

b. PACE and University of Utah (2000). 

c. Hightower et. al. (2000). 

d. BJC (2000). 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 PC-1 Effective Date: 01/31/2007 Page 11 of 24

Table 5-2a. Bounding isotopic activity concentrations for PGDP operations (pCi/pCi 

U).

Radionuclide

1 2 3 4 5 6

Pulverizer, 

ash 

handling, 

green salt 

C-410 

Pre-1983b

Converter 

Salvage 

Line 

Pre-1983b

Converter 

Salvage 

Line 

Post1983c

Tc/Np 

recovery 

Operations, 

C-400d

Balance 

of plant 

Pre1983d

Balance 

of plant 

Post1983c

Np-237 3.38E-02 1.53E+00 3.63E-02 1.67E+01 1.59E-03 3.49E-06

Pu-239/240 8.56E-01 3.08E-02 6.75E-02 4.21E+00 3.91E-05 1.86E-06

Pu-238 1.85E-01 6.67E-03 1.46E-02 9.08E-01 8.46E-06 4.01E-07

Pu-242 4.24E-05 1.53E-06 3.35E-06 2.08E-04 1.94E-09 9.20E-11

Pu-241 3.34E+01 1.21E+00 2.63E+00 1.64E+02 1.53E-03 3.22E-07

Am-241 1.48E-01 5.35E-03 1.76E-03 7.28E-01 6.79E-06 4.01E-06

Th-230 1.12E+00 1.59E-01 3.46E-01 2.16E+01 2.01E-04 9.52E-06

Tc-99 1.14E-01 1.14E-01 1.14E-01 3.63E+01 1.14E-01 1.14E-01

U-234a 4.89E-01 4.89E-01 4.89E-01 4.89E-01 4.89E-01 4.89E-01

U-235a 2.17E-02 2.17E-02 2.17E-02 2.17E-02 2.17E-02 2.17E-02

U-238a 4.89E-01 4.89E-01 4.89E-01 4.89E-01 4.89E-01 4.89E-01

U-236a 6.56E-04 6.56E-04 6.56E-04 6.56E-04 6.56E-04 6.56E-04

a. Based on uranium isotopic distribution for natural U (BJC 2000, Table 1-5). Assuming natural as 

opposed to enriched uranium provides bounding values for uranium intake estimates.

b. PACE and University of Utah (2000). 

c. Hightower et. al. (2000).

d. BJC (2000).

When determining potential intakes of TRU/99Tc by applying the isotopic ratios presented in Table 5-2 

to urinalysis data, dose reconstructors should consider that PACE and University of Utah (2000) noted 

instances where the bioassay data in the electronic database are not consistent with original 

documents. Verify urinalysis results obtained from the electronic database against the written 

records.

5.3 IN VITRO MEASUREMENT METHODS

From the start of Plant operations in 1952, samples of urine from workers involved in enrichment 

operations were analyzed for uranium.2

At PGDP, routine urine samples were typically single voids collected during the middle of the week. 

Each could have been measured for specific gravity, pH, sugar, and albumin levels, as well as for 

uranium content. During the early years, total uranium concentrations were measured using a 

calibrated fluorimeter with a detection level of 0.005 mg/L (PACE and University of Utah 2000, p. 40).

 Over time, other workers were included in the monitoring 

program. In addition, special sampling occurred in response to incidents or issues (i.e., assessments 

of the radiological impacts of TRU elements and technetium); however, those analytical methods were 

typically performed off the PGDP site, at ORNL during early periods and later at analytical services 

contractor locations. Table 5-3 lists sampling frequencies and volumes for workers assigned to 

specific buildings during various periods. In addition, the table provides default values if the specific 

location of a claimant is not available.

 

2 Fecal sampling was occasionally performed for special studies. However, a program for routine or diagnostic monitoring of 

fecal samples was never implemented.
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 12 of 24

In later years, total uranium content was assessed by kinetic phosphorescence analysis (KPA). 

Neither methodology included isotopic determinations. 

Table 5-3. In vitro measurement frequencies.

a, b

Period Facility Frequency Volume

1953 to present C-310 Every 4 weeks Spot

1953 to present C-315 Every 4 weeks Spot

1953 to present C-340 Every 4 weeks Spot

1953 to present C-400 Every 4 weeks Spot

1953 to present C-410 Every 4 weeks Spot

1953 to present Remainder Yearly Spot

1959 to present All Within 0.5 hr of 

incident

Spot

1960 C-410 Every 2 mo Spot

1960 C-340 Every 3 mo Spot

1960 C-331, C-333, 

C-335, C-337

Every 4 mo Spot

1960 C-410 Every 6 mo Spot

1991 to present All Every 4 weeks 24-hr collection or 

simulated 24-hr collection

Default N.A. Every 4 weeks Spot

a. Sources: PACE and University of Utah (2000, Section 7.4); Author unknown (no date 1).

b. There is evidence that in 1960 the measurement frequency for Building C-410 and C-340 

workers could have been either every 4 weeks or every 2 mo (C-410) or 4 mo (C-340). 

However, there is no referenceable indication whether there was a special class of worker 

to which the 2-mo frequency applied. The overlapping values have been left in the table.

5.3.1

Table 5-4 lists the in vitro measurement types and detection levels during various periods. For 

samples analyzed at the site, the detection level for total uranium in urine was reported as 0.005 mg/L 

(PACE and University of Utah 2000, Section 4.2.1). If detection levels for specific methods were not 

found in the available references, levels specified as typical in ICRP (1989) were used.

Measurement Types and Detection Levels

In addition, if it is not clear from the monitoring records how or where a particular claimant’s sample 

was analyzed, assume that they were analyzed in-house (i.e., at PGDP) and use the MDC from that 

measurement type to assess missed dose. Finally, if a record contains a notation of “less than X 

micrograms/L” or “< x pCi/sample”, that value should reflect the MDC for that sample.

5.3.2

A variety of codes occurs on various urine bioassay records for PGDP. Table 5-5 summarizes those 

known at the date of this report, along with their interpretations.

Reporting Formats and Codes

5.4 IN VIVO MEASUREMENT METHODS

Whole-body counting and other in vivo methods were implemented beginning in the early 1950s. 

Until the present (2003), this measurement method was used primarily in response to incidents, or for 

assessing the magnitude of insoluble material intakes. Even when routine whole-body counting was 

instituted for certain PGDP employees in the late 1960s, the counting frequency was sporadic and 

seldom greater than once per year.
Document No. ORAUT-TKBS-0019-5 Revision No. 01 PC-1 Effective Date: 01/31/2007 Page 13 of 24

Table 5-4. In vitro measurement types and detection levels for various periods.

a

Period Measurement type Radionuclide MDCa Comments

1952–1998 PGDP fluorimetryb Total uranium 5 μg/L

1985–1989 ORNLd U-234, -235, and -238 0.03 pCi/sample

1989–present ORNL U-234, -235, and -238 0.01 pCi/sample

1989–present Contractore U-234, -235, and -238 0.3 pCi/L

1999-Present KPAc Total uranium 0.06μg/sample DR should use the cite MDA for all urinalysis 

results by KPA, regardless of analysis date

1999–present ORNL Natural uranium 0.06 mg/sample

1969–1985 PGDP gross betac Tc-99 10 dpm/ml 10 dpm/mL

1985–1989 ORNL Isotopic plutonium 0.02 pCi/sample

1989–present ORNL Isotopic plutonium 0.01 pCi/sample

1985–1989 ORNL beta counting Tc-99 18.1 pCi/sample

1989–present ORNL Tc-99 90.9 pCi/sample

1999–present ORNL Th-228, -230, and -232 0.014 pCi/sample

1985 to 1989 ORNL Np-237 0.04 pCi/sample

1989–present ORNL Np-237 0.01 pCi/sample

1985–1989 ORNL Pu-238, -239, and -240 0.02 pCi/sample

1989–present ORNL Pu-238, -239, and -240 0.01 pCi/sample

1985–1989 ORNL Am-241 0.02 pCi/sample

1989–present ORNL Am-241 0.01 pCi/sample

a. MDC = minimum detectable concentration.

b. PACE and University of Utah (2000, p. 40) documents use of fluorimetry through 1982. No definitive reference for the specific method 

used for Total Uranium could be found in available documentation from 1983 -1998. It is favorable to the claimant to assume 

fluorimetry (as opposed to KPA) during this period. 

c. BJC (1999), Table 3.9. 

d. ORAUT (2004b). 

e. ICRP (1989). 

5.4.1

At PGDP, whole-body counting was performed using a mobile counter provided by the Y-12 Plant 

(sometimes referred to as the MMES Counter) and at other facilities. Table 5-6 lists general 

information about the detection capabilities of this counting system for various periods.

Measurement Types and Detection Levels

5.4.2

A variety of codes and reporting formats appear in the in vivo bioassay records. Table 5-7 lists the 

known codes with their interpretations.

Reporting Formats and Codes

5.4.3

On occasion, in vivo measurement results included 137Cs. However, those PGDP workers could have 

had body burdens of 137Cs from nonoccupational sources (e.g., fallout and consumption of specific 

foodstuffs). There is no evidence of occupational intakes of 137Cs at PGDP, so no dose of record 

should be associated with these measurement results, if any.

Instructions for Addressing Possible Interferences and Uncertainties

5.4.4

In general, available urine results should be considered the primary method of dose reconstruction. In 

vivo measurements, especially in the earlier years of operation, were not used for routine monitoring 

purposes. However, those results might be useful for verifying assessments of dose based on urine 

Assessment of Intake for Monitored Employees
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 14 of 24

Table 5-5. In vitro record codes.

a,b,c,d

Form identifier

Measurement 

type

Column 

identifier Code Interpretation

WCP-455 Urine bioassay Reason for Visit 33, 35 Industrial health recheck

WCP-455 Urine bioassay Reason for Visit OB, 39, 35-1 Recall sample requested following elevated 

sample

WCP-455 Urine bioassay Reason for Visit MM recall Monday morning recall sample; requested after 

days off work

WCP-455 Urine bioassay Reason for Visit 32, 33, Term Termination samples

WCP-455 Urine bioassay Reason for Visit 30, 22 Rehire

WCP-455 Urine bioassay Reason for Visit 37, 18, Per. Periodic physical; confirmatory samples were 

collected during routine physicals; this typically did 

not pertain to those on routine monitoring program.

WCP-455 Urine bioassay Reason for Visit 26 Preemployment

WCP-455 Urine bioassay Reason for Visit 005, 07, 60, 

Exposure, Special, 

Release

Samples collected following exposure or potential 

exposure in uranium release or spill

WCP-455 Urine bioassay Bottle No. Permanent sample number

PGDP_HISTORICAL_URINE Urine bioassay Results There is no distinction between positive results 

and detection limits.

PGDP_HISTORICAL_URINE Urine bioassay Sample Type Physical Routine physicals included collection of bioassay 

sample (random sampling program); this typically 

did not pertain to those on routine monitoring 

program.

WCP-455 Urine bioassay Reason for Visit I.H.R. or IHR Industrial health recheck (associated with routine 

physicals)

WCP-455 Urine bioassay Top of Card “A” Refers to shift worked (“A” = day shift)

WCP-455 Urine bioassay Top of Card “B” Refers to shift worked (“B” = evening shift)

WCP-455 Urine bioassay Top of Card “O” Refers to shift worked (“O” = midnight shift)

WCP-455 Urine bioassay F, HG and 

OTHER

“B” (128) Indicates shift and hours worked

PGDP_ANALIS_URINE Urine bioassay Results Results are given in μg/L.

5EA HPINT - Bioassay Results Report Urine bioassay Units 4 dpm/L

5EA HPINT - Bioassay Results Report Urine bioassay Units 1 dpm/ml

5EA HPINT - Bioassay Results Report Urine bioassay Units 2 dpm/day

5EA HPINT - Bioassay Results Report Urine bioassay Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Urine bioassay Units 5 μg/ml

5EA HPINT - Bioassay Results Report Urine bioassay Units 6 Bq/L

5EA HPINT - Bioassay Results Report Urine bioassay Units 7 Bq/day

5EA HPINT - Bioassay Results Report Urine bioassay Reason 3 Routine

5EA HPINT - Bioassay Results Report Fecal analysis Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Fecal analysis Units 7 Bq/day
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 15 of 24

Form identifier

Measurement 

type

Column 

identifier Code Interpretation

5EA HPINT - Bioassay Results Report Breath analysis Units 2 dpm/day

5EA HPINT - Bioassay Results Report Breath analysis Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Breath analysis Units 4 dpm/L

5EA HPINT - Bioassay Results Report Breath analysis Units 6 Bq/L

5EA HPINT - Bioassay Results Report Breath analysis Units 7 Bq/day

WCP-885 Urine bioassay Schedule A Day sample is to be taken (1=Monday; 

2=Tuesday; 3=Wednesday; 4=Thursday)

WCP-885 Urine bioassay Schedule B Type analysis (1=Uranium; 2=Fluoride; 

3=Mercury)

WCP-885 Urine bioassay Location Shifts A, B, O and D

Frequency Uranium #1 = 1 a month; Fluoride #2 = 1 a year; 

Mercury #3 = Blank

Permanent log (sample collection log) Urine bioassay Number Permanent sample 

number

Numbers are consecutive from top to bottom of 

page and continue from one page to next.

IBM Report Cards Urine bioassay A Name

IBM Report Cards Urine bioassay B Badge

IBM Report Cards Urine bioassay C Date (date shown on log)

IBM Report Cards Urine bioassay D Code for this service

IBM Report Cards Urine bioassay E Analysis results in boxes labeled Uranium, 

Fluorides, and Mercury

UCN-5242, “Sample Analysis (Medical).” Urine bioassay Same as for 

NCP-455

Same as for NCP455This form number was in use from 1969 to 1970.

a. Sources: Maisler (2003); Eckerman and Ward (1992); Author unknown (no date 1). 

b. Form WCP-885 is referred to as “NCP-885” in Eckerman and Ward (1992); however, they appear to identify the same form.

c. Around June 1956, form WCP-455 was modified with additional columns added. Individual bioassay records reviewed indicates the exact date the newly expanded 

form was used. This varies a little from person to person, but was in the middle of 1956 for all records reviewed. 

d. The bioassay records we have starting in 1977 (exact dates vary from person to person) are not copies of results recorded on bioassay cards; they are database 

printouts (e.g., “PGDP_Historical_Urine”). The uranium results are in units of μg/L with results recorded to the nearest integer, i.e., 0, 1, 2, etc. 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 16 of 24

Table 5-6. In vivo measurement types and detection levels for various periods. 

Period Equipment

Measurement 

type Radionuclide

MDAa

(units of 

record)

Action level 

for recount

Action level 

for work 

restriction

1958 ORNLb Lung Pu-238 80 nCi Not specified Not specified

1960–1967 Y-12c Whole body Np-237 2.7 nCi Not specified Not specified

1968–1980 Y-12 mobile 

counterd 

Whole body U-235 0.083 mg Not specified Not specified

1968–1980 Y-12 mobile 

counterd

Whole body U-238 4 mg Not specified Not specified

1968–1980 Y-12 mobile 

counterd

Whole body Np-237 17,000 pCi Not specified Not specified

1965–1991 Y-12 mobile 

counterd Lung Total uranium 4 mg 4 mg 27 mg

1965–1991 Y-12 mobile 

counterd

Lung Enriched 

uranium (2% 235U)

0.1 mg 0.1 mg 0.24 mg

1965–1991 Y-12 mobile 

counterd

Lung Depleted 

uranium

4 mg 4 mg 37 mg

1965–1991 Y-12 mobile 

countere 

Lung Np-237 200 pCi 1,700 pCi 17,000 pCi

1991–1995 Helgeson counterd Lung Total uranium 2 – 4 mg 2 – 4 mg 27 mg

1991–1995 Helgeson counterd Lung Enriched 

uranium

0.04 – 0.07 

mg

0.04 – 0.07 

mg

0.24 mg

>1995 No counting performed

a. MDA = minimum detectable activity or amount. The MDAs shown for uranium, while given in units of mass, are 

presumed to have been based on measurement of Th-234 activities along with an assumed isotopic ratio. It is 

presumed that the results for enriched uranium are based on measurement of U-235 activities and an assumption of 

enrichment. However, these presumptions cannot be confirmed so cautious use of these MDA values is recommended.

b. ORAUT (2004b). 

c. ORAUT (2006). 

d. ORAUT (2004c). 

e. Hill and Strom (1993).

bioassay results, in determining likely absorption types, or in providing upper and lower limits to the 

range of possible doses. 

5.5 SIGNIFICANT INCIDENTS WITH INTERNAL DOSE POTENTIAL

During operations at PGDP, a number of incidents occurred that increased the potential for intakes of 

radioactive materials. If a claimant (or employee) recalls involvement in one or more of those 

incidents, dose reconstructors can use the information in Table 5-8 to identify the dates, location, 

and/or source term for an incident-specific dose assessment.
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 17 of 24

Table 5-7. In vivo record codes.

a

Form identifier

Measurement 

type

Column 

identifier Code Interpretation

In Vivo Radiation 

Monitoring Report

Whole-body 

count

Surface 

contamination 

Check mark, 

yes

Surface contamination on the 

subject was checked. If 

measurable activity was 

found, it was so noted on 

card.

In Vivo Radiation 

Monitoring Report

Whole-body 

count

Analysis 

Sequence

These align with Output -

Analysis Sequence Results 

listed at bottom of card; they 

are not relevant to dose 

reconstruction process.

In Vivo Radiation 

Monitoring Report

Whole-body 

count

Output - Analysis 

Sequence 

Results

A. Enriched 

Uranium

The maximum U-235 

enrichment was 2% until 

2000, after which it was 5%.

In Vivo Radiation 

Monitoring Report

Whole-body 

count

Output - Analysis 

Sequence 

Results

J. NLO 

Uranium

Refers to special spectrum 

region of interest for National 

Lead of Ohio, early operator 

of Fernald facility.

In Vivo Radiation 

Monitoring Report

Whole-body 

count

Hand-written 

notes

No Np noted Presence or absence of Np237 was qualitatively 

evaluated.

In Vivo Radiation 

Monitoring Report (with 

boxes for data entry)

Whole-body 

count

Hand-written 

notes

BFD Initials of whole-body counter 

operator (individual who filled 

out the card)

5EA HPINT - Bioassay 

Results Report

In vivo records Units M μCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units N nCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units P pCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units D dpm

5EA HPINT - Bioassay 

Results Report

In vivo records Units B Bq

5EA HPINT - Bioassay 

Results Report

In vivo records Units U μg

a Sources: Maisler (2003); Eckerman and Ward (1992); Author unknown (no date 1). 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 18 of 24

Table 5-8. A history of significant incidents and events.

a

Incident date Incident description Facility Other information

1952–1990 Exposure to UF4, UO, and process dust 

during guard patrolling

All buildings

1952–1980 Exposure to uranium metal

July 1953 First use of reactor tails

November 1956 Fire C-310

1957–1977 Green salt, black oxides on floors and other 

surfaces

C-340

December 1962 Fire C-337

March 1962 Explosion and fire C-340 One fatality

April 1968 Worker overexposure Unknown Two workers overexposed

January 1978 Fire C-315

1958 to 1962 Cascade improvement program C-331, C-333, 

C-335, and C-337

1974 to 1982 Cascade improvement program C-331, C-333, 

C-335, and C-337

Neptunium production C-400

1980 to 1982 Exposure to UF4 and uranium dust during 

drum crushing

C-746

a. Source: PACE and University of Utah (2000).
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 19 of 24

REFERENCES

Author unknown, no date 1, “Urinalysis Program,” from Paducah Gaseous Diffusion Plant files, 

Paducah, Kentucky. [SRDB Ref ID: 12651]

Author unknown, no date 2, “Uranium in Urine Analysis,” from Paducah Gaseous Diffusion Plant files, 

Paducah, Kentucky. [SRDB Ref ID: 23001]

Baker, R. C., 1987, “Exposure Assessment - Uranium Recycle Materials in the Paducah Feed Plant,” 

Attachment 2 of Bostock, D. J., 1987, "Exposure Assessment, Paducah Feed Plant," letter to 

J. Parks (U.S. Department of Energy), Martin Marietta Energy Systems, Paducah, Kentucky, 

July 23. [SRDB Ref ID: 3902]

Barton, C. B., 2006, "Verifying and Expanding Paducah Occupational Internal Dose Document 

Tables," memorandum to D. S. Mantooth, Dade Moeller & Associates, Richland, Washington, 

May.BJC (Bechtel Jacobs Company) 1999, Internal Dosimtery Technical asis Document for 

Bechtel Jacobs Company LLC, BJC/OR-258. [SRDB Ref ID: 12678]

BJC (Bechtel Jacobs Company), 2000, Recycled Uranium Mass Balance Project Paducah Gaseous 

Diffusion Plant Site Report, Draft for Review, BJC/PGDP-167, Paducah Gaseous Diffusion 

Plant, Paducah, Kentucky, March 5. [SRDB Ref ID: 16498]

DOE (U. S. Department of Energy), 2000, Phase II, Independent Investigation of the Paducah 

Gaseous Diffusion Plant, Environment, Safety and Health Practices, 1952-1990, Office of 

Oversight, Environment, Safety and Health, Washington, D.C., February. [SRDB Ref ID: 

12040]

Eckerman, K. F., and R. C. Ward, 1992, DOSEXPRT - A Bioassay Dosimetry Code for Martin 

Marietta Energy Systems, Inc., ORNL/TM-11857, Martin Marietta Energy Systems, Oak Ridge 

National Laboratory, Oak Ridge, Tennessee, April. [SRDB Ref ID: 12548]

Hightower, J. R., L. R. Dole, D. W. Lee, G. E. Michaels, M. I. Morris, D. G. O'Conner, S. J. Pawel, 

R. A. Schmoyer, L. D. Trowbridge, and V. S. White, 2000, Strategy for Characterizing 

Transuranics and Technetium Contamination in Depleted UF6 Cylinders, ORNL/TM-2000/242, 

UT-Battelle, Oak Ridge National Laboratory, Oak Ridge, Tennessee, October. [SRDB Ref ID: 

23571]

Hill, R. L., and D. J. Strom, 1993, Internal Dosimetry Technical Basis Manual for Portsmouth and 

Paducah Gaseous Diffusion Plants, PNL-8723, Battelle Memorial Institute, Pacific Northwest 

Laboratory, Richland, Washington, June. [SRDB Ref ID: 12024]

ICRP (International Commission on Radiological Protection), 1989, Individual Monitoring for Intakes of 

Radionuclides by Workers: Design and Interpretation, Publication 54, Pergamon Press, 

Oxford, England.

ITC (International Technology Corporation), 1992, Personnel Exposure Potential to Transuranic 

Materials at the Paducah Gaseous Diffusion Plant, IT/NS-91-109, Washington, D.C., 

September 2. [SRDB Ref ID: 12682]

Maisler, J., 2003, “Results of Onsite Interview at Paducah," memorandum to C. Berger, Integrated 

Environmental Management, Dayton, Ohio, November 20. [SRDB Ref ID: 13764]
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 20 of 24

NCRP (National Council on Radiation Protection and Measurements), 1987, Use of Bioassay 

Procedures for Assessment of Internal Radionuclide Deposition, Report 87, Bethesda, 

Maryland.

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Basis Document for the 

Paducah Gaseous Diffusion Plant – Site Description, ORAUT-TKBS-0019-2, Revision 00, Oak 

Ridge, Tennessee, May 7.

ORAUT (Oak Ridge Associated Universities Team), 2004b, Technical Basis Document for the Oak 

Ridge National Laboratory – Occupational Internal Dose, ORAUT-TKBS-0012-5, Oak Ridge, 

Tennessee, June 29.

ORAUT (Oak Ridge Associated Universities Team), 2004c, Technical Basis Document for Portsmouth 

Gaseous Diffusion Plant – Occupational Internal Dose, ORAUT-TKBS-0015-5, Rev. 00, Oak 

Ridge, Tennessee, November 24.

ORAUT (Oak Ridge Associated Universities Team), 2005, Dose Reconstruction Considerations for 

Recycled Uranium Contaminants, ORAUT-OTIB-0053, draft, Oak Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2006, Y-12 National Security Complex – Internal 

Dose, ORAUT-TKBS-0014-5, Rev. 01 PC-3, Oak Ridge, Tennessee.

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, 

December. [SRDB Ref ID: 10870]

PGDP (Paducah Gaseous Diffusion Plant), 1991, PGDP Monthly Urinalysis Results, March 1989 to 

March 1991, Paducah, Kentucky.

Smith, R. F., 1984, Historical Impact of Reactor Tails on the Paducah Cascade, KY/L-1239, Union 

Carbide Corporation, Paducah Gaseous Diffusion Plant, Paducah, Kentucky, March 19. 

[SRDB Ref ID: 16826]
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 21 of 24

GLOSSARY

absorption type 

Categories for materials according to their speed of absorption in the body, which replaced the 

earlier inhalation clearance classes. Defined by the International Commission on Radiological 

Protection, the absorption types are F for fast absorption (formerly inhalation class D), M for 

moderate absorption (formerly inhalation class W), and S for slow absorption (formerly 

inhalation class Y). Also called solubility type.

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol. In relation to health 

physics, normally assumed to be 5 micrometers.

bioassay

Determination of kinds, quantities, or concentrations, and in some cases locations of 

radioactive material in a living body, whether by direct measurement (in vivo measurement) or 

by analysis and evaluation of materials excreted or removed from the body (in vitro

measurement).

body burden

Amount of radioactive material in an individual's body at a particular point in time.

dose 

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

• Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

• Collective dose is the sum of the doses to a specific population.

• Committed dose is the dose over time (usually 50 years for workers) to a specific organ 

or tissue from an intake of radioactive material.

• Cumulative dose is the sum of all doses to the same portion of the body or to the whole 

body over time.

• Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).

• Effective dose is the sum of the equivalent doses in the principal tissues and organs of 

the body, each weighted by a tissue weighting factor that accounts for the probabilities 

of fatal and nonfatal cancers according to severity and the average length of life lost 

due to an induced cancer. It indicates the biological effect of the radiation exposure in 

that tissue. 

• Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied 

by a weighting factor for the particular type of radiation. 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 22 of 24

• Organ dose is the dose to a specific organ.

• Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation. 

• Personal dose equivalent is the dose equivalent in soft tissue below a specified point 

on the body at a specified depth. 

• Shallow dose is the dose at a 0.07-centimeter depth in tissue (7 milligrams per square 

centimeter).

• Skin dose is the dose to the skin.

• Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.

dose equivalent (H, DE) 

Product of absorbed dose in units of rem or sievert in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens. In this document, exposure does not refer to the radiological physics concept of 

charge liberated per unit mass of air.

insoluble material

A term loosely used to describe the relative degree of solubility of a material in body fluids. 

Recognizing that no material is absolutely insoluble, the terms low solubility or poorly soluble 

are preferable.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds. Depending on the radionuclide involved, intakes are in units of 

mass, activity, or potential alpha energy.

internal dose or exposure

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent internal radiation dose 

based on measurements in the work environment and/or bioassay. 

in vitro

Of or relating to a process that takes place under artificial conditions or outside a living 

organism (e.g., in the laboratory). From Latin meaning in glass. 

in vivo

Of or relating to a process that takes place in a living organism. From Latin meaning in life. 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 23 of 24

lung solubility type 

See absorption type. 

minimum detectable amount (MDA)

Lowest amount of radioactive activity or substance amount detectable by a specific instrument 

or process. Smallest amount or activity of a radionuclide in a sample or organ that yields a 

result above the detection level with a specific probability of a Type II (false negative) error 

while accepting an specific probability of a Type I (false positive) error. 

minimum detectable concentration (MDC)

Lowest concentration of a material in a substance (e.g., urine) detectable by a specific 

instrument or process. Minimum detectable activity (or amount) in units of concentration. 

minimum reporting level 

Level below which an analytical dose is not recorded in the worker’s dose record, usually 

based on a site-specific policy decision. The recording level is not necessarily the same as 

the minimum detectable amount or activity for that measurement. Also called less-than value, 

minimum reportable dose, minimum recordable or recording dose, recording level, and 

reporting level. 

monitoring (personnel)

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements. 

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment.

radiation

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body. 

Radiation, as used in this document, does not include nonionizing radiation, such as radio- or 

microwaves, or visible, infrared, or ultraviolet light.

reactor tails 

Recycled uranium (typically UO3) from reactor operations (typically Savannah River and 

Hanford) that contains traces of transuranic isotopes not removed during chemical processing 

(e.g., reduction-oxidation). 

recording level

See minimum reporting level. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 
Document No. ORAUT-TKBS-0019-5 Revision No. 01 Effective Date: 10/25/2006 Page 24 of 24

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural. 

routine monitoring

Monitoring carried out at regular intervals during normal operations.

sievert (Sv) 

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

special monitoring

Monitoring in addition to the routine monitoring program carried out for special purposes such 

as estimating the amount of radionuclide deposited in a person after a known or suspected 

accidental intake or after a known or suspected environmental release. 

spot sample

In relation to bioassay, usually a single void of urine.